A helium coolant side-stream purification system consisting of parallel sections for radioactive and non-radioactive de-contamination was designed for the proposed Pebble Bed Reactor Experiment. Primary equipment components are two gas coolers, gas heater, charcoal delay trap, CuO oxidizer, Molecular Sieve adsorber, and full flow filter. The charcoal delay trap is sized to provide a hold-up of 30 minutes for Kr isotopes, 6hr hold-up for Xe isotopes, and 99.9% retention of iodine isotopes resulting in "de-contamination factors" varying from l for Kr85 to 556 for I131. Non-radioactive de-contamination will result in a steady state concentration of CO2 in the coolant of 20.8ppm or less.

A correlation between the optical densities if the Pb-Sn alloy system and UO2 with respect to Co^60 gamma radiation has been determined. This enables one to fabricate density standards of whatever geometry may be desired for one in the gamma absorptiometer by simply casting a Pb-Sn alloy of the proper composition to correspond to the density required.

The primary purpose of a cleanup system in a recirculating water loop is to maintain the best possible water quality conditions. This is normally accomplished by continuously purifying all or a portion of the coolant. A secondary objective of the cleanup system is to help maintain the system pH at a constant value. A system that will satisfactorily accomplish both of these objectives is at times difficult to obtain. Generally the pH control characteristics are sacrificed in favor of the more important cleanup requirements. A somewhat new approach to the problem pf cleanup system design appears to offer a solution to this problem for high pH systems.

During normal operation the NPR will function as a closed system and the coolant will be maintained at a high degree of purity. Periodically, however, the system will have to be opened for fuel element discharge and/or decontamination. During these periods the reactor will be operated on single pass cooling. The use of deionized and deoxygenated water as the single pass coolant is not economically feasible due to the large quantities of coolant that will be required. At present it appears that filtered water will be used for this purpose. Since filtered water contains substantial amounts of dissolved solids (approximately 100 parts per million), and since it is saturated with air, this type of operation will have to be investigated with regard to corrosion problems. A particularly serious problem will be present after the system has been decontaminated, since the high temperature oxide films will have been removed and the bars metal surfaces will be exposed to the coolant.

In some engineering applications it has become necessary to operate equipment containing small diameter rods or tubes oriented parallel to flow stream. In the case of several nuclear reactors such as the Plutonium Recycle Test Reactor and the S. S. Savannah Maritime Reactor, bundles of small diameter rods are used as the fuel elements. The situation also has application to heat exchangers. A method for accurately predicting the pressure drop characteristics of various bundle configurations would be helpful in the design and selection of equipment. Some progress has been made toward gaining a greater knowledge of the bundle characteristics under low temperature conditions. This report includes the development of relationship for the effect of temperature on pressure drop.

The oscillator of the 88-in. cyclotron which is being built in Berkeley is tunable from 5.3 to 16.5 Mc. It delivers a maximum c-w power of 300 kw. At the rated doc voltage of 75 kv the resonator stores 4.5 joules of electrical energy. The transients produced by this amount of energy, during sparking, place unusual requirements upon the design of the oscillator tube. The features of the RCA 6949 which make it particularly well-suited to this type of application are discussed in this paper. Other topics covered are the oscillator anode power supply, the hard-tube modulator, protective equipment, and oscillator instrumentation.

The probabilities for nine electronic transitions among the low lying excited states in the C2 molecule are calculated by the dipole moment operator method and are given in the form of oscillator strength (or f values).

The concept of thermonuclear explosives as a potentially cheap and almost inexhaustible energy source for mankind's non military needs has for several years been under active consideration at the Lawrence Radiation Laboratory. Many of the proposed peaceful applications involve underground nuclear explosions, and several experiments at the AEC Nevada Test Site have provided valuable insight into the phenomenology of such explosions. Among the possible uses currently under consideration are excavation, heat production, isotope production, mining, recovery of oil from shales and tar sands, improvements of ground water supplies, and the construction of earth fill dams. In addition a program of experimental research in the laboratory and in the field is under way. Sometime in 1961 Project Gnome if approved will be conducted in New Mexico. The purpose of Gnome, a contained nuclear explosion in a salt deposit, is to study the feasibility of heat recovery and isotope production, neutron scattering experiments will also be included. Other proposed nuclear projects will involve the creation of a small harbor near Cape Thompson, Alaska as the result of an experiment designed to investigate the cratering effects of nuclear explosives; a proposal to investigate the recovery of oil from Canadian tar sands using thermonuclear …

Reactivity measurements made by the pulsed neutron technique were compared with results obtained by conventional techniques. The pulsed neutron results were in good agreement with those obtained by stable period measurement and rod drop. Differential effectiveness of partially inserted rods was shown to be well represented by elementary perturbation theory. Finally, the pulsed neutron technique was found to be the only good method for measurement of large reactivity changes.

Vanadium group beryllides were prepared and studied by x - ray powder diffraction and crystallographic methods. Properties included phase studies, sintering studies, vapor pressure measurements, and thermal diffusivity measurements,

Summary: Prototype testing of the safety and control rod drives indicated that both units functioned properly. No major problems were encountered during testing. Seal leakage data collected indicated that the seal units were performing satisfactorily. Scram times during both cold and hot testing were excellent and actually better than expected.

The objective of this test is to measure the emergency reactor flow through the high pressure crosstie line (HPCT) after the removal of the flow limiting orifices in the HPCT. The flow limiting orifices in the HPCT were removed on July 5, 1960, as per Design Change No. 383. The removal of the flow limiting orifice allowed increased emergency flow and brought the crosstie coolant flow more nearly into conformance with the coolant supply reliability criteria. The purpose of this test is to measure emergency flow under certain conditions so that available flow under all conditions may be more precisely determined.

It is recommended that the ``square pile`` concept be adopted for all disaster total control calculations, and that the basic reactor constants listed in HW-62884, except for Ball 3X local strength at the DR Reactor, be used in applying this method. Curves are included for each reactor type, indicating allowable enrichment based on appropriate local control strengths. (The reactors whose operating methods are affected by disaster total control requirements are B, D, F, and DR Reactors; the remaining piles have sufficient geometrical coverage). An example of the analytical method is included.

The objective of this irradiation is to further verify the corrosion rate of tubular-type fuel elements under conditions of high specific power and central core temperatures. This fuel will be the inner tube only of an NPR fuel assembly. As in previous tests, this inner tube rupture will be used to further substantiate the rupture detection instrumentation that is being used in the development of the NPR. Previously unirradiated fuel will be used in this test. The reactor is to operate at full power during the test. Permission is requested for charging two tubular elements The top element will have attached to it a hydraulic mechanism for opening a defect in the outer surface of the tube. The second or bottom element, will be used as a heater element to maintain loop temperature.

Preliminary engineering evaluations indicate that piping in the 105 B, D, F, DR, and H reactors has deteriorated to the extent that an increasing rate of component failure can be expected. In view of this, a budget submission was made in the FY-1962 P. A. and C budget and has been included in the I.P.D. Plant Improvement Program. The purpose of this report is to substantiate the need for this program and to review information generated during the past three years concerning the condition of rear face piping and hardware. This review includes the history of rear face piping and hardware problems, study activities undertaken to date to ascertain the condition of the components, action taken to correct actual component failures, programs recommended to correct deficiencies which operating experience and engineering judgement indicate are necessary, and programs to accumulate additional information to support design of new piping and hardware components.

Steady state hydraulic demand curves were obtained for tube powers of 500, 1000, 1500 and 2000 KW with an inlet water temperature of 20C and a rear header pressure of 25 psig. These curves are shown in figures. The point of initial unstable flow for various tube powers is shown for a front header pressure of 325 psig. The flow rate that would lead to the initial point of unstable flow as a result of a sudden plug upstream of the Panellit tap is shown in a figure.

The neutrons in an operating nuclear reactor are generated primarily by the fission events which are taking place. The great bulk of these relatively high energy or fast neutrons are slowed down or thermalized by a series of elastic collisions with the moderator nuclei which comprise the bulk of the volume of the reactor core. Once slowed down, the neutrons diffuse through the reactor core until they are absorbed or eliminated by some other process. Each of these three groups of neutrons, i.e., the fast or source neutrons, the intermediate or slowing down neutrons, and the slow or thermal neutrons, has a characteristic energy distribution. At a constant power level or rate of fissioning, an equilibrium is soon established among the groups at any point in the reactor. If it is assumed that a smooth transition exists between the different energy groups, it is possible to evaluate the entire neutron spectrum at a point in the reactor by determining the parameters which characterize each of the three groups. This has been done in the F Reactor Quickie Facility using radioactivants.

The following report describes the injection and trapping of high current electron beams in order to construct an electron gun and the first 2 Mev section of the accelerator.

A study was made to determine the coolant flow required to prevent the temperature of the graphite moderator and reflector blocks from exceeding 1100 deg F at full-power operating conditions. The temperature distribution in the graphite blocks was also determined. Tha moderator was primarily cooled by coolant flowing in the 1/8-inch annulus formed by the fuel-assembly sleeves and the fuel-channel holes in the moderator blocks. Coolant flow in the controlrod channels and in the experimental-tube coolant annuli also cooled the moderator, and this effect was considered in setting moderator-annulus coolant flow. The coolant flows required for each of four zones in the reactor core were determined. The total moderator-annulus coolant flow (excluding 16 unfueled channels) was 15,200 lb/hr. The moderator-block temperature distributions for fullpower reactor operation after 20 years of operation are given. The maximum temperature (1100 deg F average over the cross section) occurred near the top of the graphite block. The temperature in the majority of the graphite columns varied from 600 to 1000 deg F over the lower half of the column and from 1000 to 1100 deg F over the top half of the column. The maximum graphite surface temperature was less than 1100 deg F. …

During the report period, the nature of the ORNL supporting activities gradually changed, reflecting the progress in the conetruction of the N.S. Savannah. Design reviews continued to require considerable attention. The Inspection Engineering Dept. of the Jab. continued to act as an inspection agency for the ALC, witnessing inspections and tests during fabrication of components of the nuclear reactor system. An enviromnental analysis was made of safeguard aspects of operation of the N.S. Savannah at the NYSC site in Cannden. A study of the safety response of the reactor on the ORNL Analog Computer further defined the important role of the Doppler coefficient in controlling reactivity excursions. Health physics aspects of the operation were studied. Limited waste disposal studies indicated that sea disposal of exhausted demineralizer resins may be facilitated by casting the radioactive resins into concrete. Installation of a pressurized-water in-pile test loop in the ORR neared completion. The neutron flux distribution in the loop was determined by experimental measurements in a nuclear mockup of the in-pile section. Metallurgical activities included nil-ductility testing of steel from the reactor vessel, chemical analyses of primary system components, and investigations of the properties of electroless - nickel brazed joints. Limited fabrication studies …

The formation of heavy negative ions by the attachment of low-energy electrons to oxygen molecules was studied for small amounts of oxygen mixed with methane or carbon dioxide. The rate of attachment in both cases was found to depend on the electron energy, the pressure of the oxygen and the non-attaching gas, and on the kind of non-attaching gas. In general, the attachment increases as electron enprgy decreases or as either oxygen or total pressure increases. The value of the attachment coefficient in oxygencarbon dioxide mixtures is about 100 times its value in oxygen-methane mixtures. This large difference is probably due in part to differences in electron energy and partly to differences in the stabilizing qualities of the two molecules. Dissociative attachment, which should be pressure independent, does not occur at the low energies that were used in this work. Both methane and carbon dioxide are to differences in the stabilizing qualities of the two molecules. Dissociative attachment, which should be pressure independent, does not occur at the low energies that were used in this work. Both methane and carbon dioxide are sometimes used as filling gases for Geiger and proportional counters. The high sensitivity of carbon dioxide to oxygen …

Lengths of pistillate inflorescences of sedges (Carex spp.) growing in the contaminated soils of White Oak Lake bed were measured in relation to radiation fields. Carex Frankii Kunth and Carex vulpinoidea Michx. populations were sampled from areas with air dose rates of 0, 10, 20, and 40 mr/hr. Analyses of variance showed that mean lengths of inflorescences of C. vulpinoidea were simllar to each other in these areas but that those of C. Frankii were significantly different. However, the sites differ in such factors as soil moisture, fertility, and alkalinity, so that these differences in the length of inflorescence may not be accounted for primarily by the exposure dose rate of the radiation field. (auth)

The saturation voltage and current characteristics of a compensated ionization chamber (Q-1045) were measured with special regard to high current and voltage ranges. The chamber can be operated at currents up to I ma with a 2000 volt power supply. (auth)

Corrosion specimens of the five major aluminum alloys used in the construction of the Oak Ridge Research Reactor have been exposed to the high- purity primary cooling water in the ORR core and in the extennal portion of the primary cooling loop to determine their corrosion rates under actual operating conditions. These alloys, 1100, 3003, 5052, 5154. and 8061, exhibited average corrosion rates of less than 2.6 mpy during the first 500-hr test period and less than 0.5 mpy for a 4032-hr test. Very superficial pitting was observed. and no case of intergranular corrosion was found. (auth)